Pressurized-water reactors (PWR) in nuclear power plants employ fuel claddings made of zirconium alloys, of which Zircaloy 4 having the following composition is typical:
1.2-1.7% Sn;
0.18-0.24% Fe;
0.07-0.13% Cr; and
the balance being zirconium and incidental impurities (the percentage being based on weight as hereinafter).
With a view to improving the economy of the operation of nuclear power plants, efforts are being made to maximize the efficiency of fuel burnup and this has led to the need for fuel claddings to stay in the reactor for a longer period. But this need cannot be met by the conventional fuel claddings made of zirconium alloys typically Zircaloy 4, because they do not have sufficient corrosion resistance to withstand prolonged exposure to the atmosphere in the reactor.
Under these circumstances, the present inventors conducted studies in order to develop a zirconium alloy that exhibits superior corrosion resistance when used as a nuclear reactor fuel cladding material, the studies being particularly addressed to modification of Zircaloy 4. As a result, the present inventors found that an alloy additionally containing Ta, and optionally Nb, too, in the composition specified above had such improved corrosion resistance that it was suitable for use as a nuclear reactor fuel cladding material over a prolonged period.